An Acoustically Driven Magnetized Target Fusion Reactor

We propose a new acoustic compression scheme for a MTF power plant. A strong acoustic wave is produced by piston impacts. The wave focuses in liquid PbLi to compress a pre-formed FRC plasma. Simulations indicate the possibility of building an economical 60 MWe power plant. A proof-of-principle experiment produces a small D-D fusion yield of 2000 neutrons per shot.     

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the ⁹⁹TC and ¹²⁹I and ¹³⁵Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of ⁶Li in lithium (Li) and the theoretical density (TD) of Li₂O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ≥ 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.     

聚变驱动乏燃料焚烧堆(FDS-SFB)燃料循环动态分析

针对聚变驱动乏燃料焚烧堆FDS-SFB燃料循环系统与一次通过燃料循环系统,利用系统动力学软件Vensim分别建立了这两种循环系统的动态分析模型,并根据假设的三种核电发展情景,分别计算了这两种燃料循环系统的资源需求、乏燃料累积量、钚累积量及次锕系元素累积量。初步计算结果表明:与一次通过式燃料循环系统相比,FDS-SFB燃料循环系统可减少天然铀需求量与乏燃料累积量,减少的程度与核电发展规模相关。     

JENDL-4.0 Benchmarking for Fission Reactor Applications

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.     

CFBR-Ⅱ堆裂变率计算

为研究CFBR-II堆裂变率的具体分布,建立了基于MCNP程序的裂变率计算方法,并利用国外高富集度铀裸球装置进行了验算.针对CFBR-II堆实际结构的限制,设定了上下半球45o方向的虚拟孔道,沿此方向计算了裂变率的细致分布.结果表明,裂变率最大处位于高富集度铀区域中间某点,而非位于系统中心或上下半球球心.采用分布积分法进一步得到系统中心归一后的总裂变率为3 150.3,其中,贫化铀反射层贡献占5%.     

D-T Beam Fusion Reactor

The Beam Fusion Reactor (BFR) is based on a field-reversed configuration and confined ion energies in the range of hundreds of keV. Repetitively pulsed, intense ion beams sustain the ion distributions and provide current drive. In the BFR the ion orbit size is comparable to the dimensions of the confined plasma and the expectation is for classical transport of the particles and energy. Based on technologies that readily exist, or nearly so, a D-T fueled BFR could be assembled in a compact configuration that is scaleable in terms of its output energy as well as to the advanced fuel regime. In the simplest case the mean azimuthal velocities and temperatures of the two ion (fuel) species are equal and the plasma current is unneutralized by electrons; the resulting distribution functions are thermal in a moving frame of reference. Reactor kinetic calculations are based on the Vlasov-Maxwell equation, including a Fokker Planck collision operator and all sources and sinks for energy and particle flow. The quality factor for this system is projected to be: Q P _fusion /P _{bremsstrahlung} = 104.     

Beam Driven Tokamak Fusion Reactors

The recent successes in neutral beam current drive experiments on large tokamaks prompt us to consider the prospect of a beam driven steady state tokamak fusion reactor. A neo-classical theory on the beam driven current, which fully includes the toroidal effects on the induced ion and electron currents and is the most reliable to date, is reviewed in this article. The prospect of steady state tokamaks predicted by the theory may be somewhat pessimistic if the plasma current is sustained only by the beam driven current, because too much beam power is necessary for the current drive. However, thanks to the bootstrap current which was discovered in many tokamaks in recent years, the feasibility of a continuous tokamak operation will probably be highly improved because several tens of percent of the current will be produced by the bootstrap current. The energy gain Q would attain 30 to 50 in the power reactor and about 10 even for next generation (demonstration size) tokamaks.     

US Fusion Reactor Technology Program

An outline of the current status and future strategy of the US Fusion Reactor Technology Program is presented.     

Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

Selection of lithium containing materials is very important in the design of a deuterium–tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid–solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.     

Can Fusion and Fission Breeding Help Civilization Survive?

As apparent from the title, this author feels that civilization faces a real threat, one which will become obvious and serious within the lifetimes of many readers of this article. This threat is not global warming, but lack of affordable energy. We take for granted turning on a light, or adjusting our thermostats in winter or summer, or filling our cars gas tank; and lose sight of the fact that there are huge and complicated industrial systems which make this possible. But as we run out of petroleum and natural gas, and worry about the environmental and climatic effects of burning coal on the required scale, how can this continue? This paper makes the case that breeding nuclear fuel, by both fusion and fission, is the only way our civilization as we know it, can continue beyond the next half century or so.     

Thinking in Different Ways to Combine Fusion with Fission

The common goal of CTR, but in particular of ICF, is low yield-high gain. Fission triggered large TN explosive devices meet the second but not the first of these conditions. These devices depend on the rare isotopes U235, Pu239, or U233, but for them the fusion energy output greatly exceeds the output from fission, limiting the fallout. In thinking about different ways to combine fusion with fission, there are three questions: (1) Are there ways where both conditions can be met, and where the fallout from fission is small? (2) Can the conditions be met without the use of U235, Pu239, or U233, but with U238, Th232, and perhaps with the fission of light nuclei like B10 or Li6, the latter having no fallout? (3) Are there concepts for MF, combining fusion with fission, without U235, Pu239 or U233? In my talk I will present reasons why under the above stated conditions two things seem to be possible: (1) The greatly facilitated fast ignition of thermonuclear microexplosions with a small amount of U238 or Th232. (2) The greatly enhanced pulsed MF burn aided by the fission of light nuclei such as B10, but also of the U238 and Th232 and with a neutron moderator. In either one of these cases the burn is “autocatalytic” in the sense that neutron-induced nuclear reactions in a halo surrounding the fusion plasma drive thermomagnetic currents compressing and increasing its neutron production rate.     

应用溶液堆生产裂变99Mo

裂变^99Mo的供应对保障核医学的应用和发展具有非常重要的作用。应用溶液堆生产裂变^99Mo具有运行安全、没有靶件制备、溶解工艺和生产工艺相对较简单、消耗铀燃料较少的优点。因此应用溶液堆生产裂变^99Mo具有很好的应用前景。     

球床包层聚变—裂变混合堆热工安全分析

球床包层混合堆与板状元件包层混合堆相比较，前者在核燃料生产和安全方面可能具有更多的优越性。本应用ＴＨＥＲＭＩＸ程序和辅助程序对我国开发的托卡马克堆芯氮气冷却球床包层聚变－裂变合堆的包层进行了热工计算。计算中考虑了不同的燃料球材料及稳态，卸压和断流事故工况。计算结果表明，只要选用合适的燃料球材料和设置适当的控制保护系统，具有快速卸料罐的托卡马克堆芯氦气包层聚变－裂变混合堆的概念设计在安全上的可行的。     

Dust Divertor for a Tokamak Fusion Reactor

Micron-size tungsten particulates find their equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic ion flow drag parallel to the divertor surface. The natural circulation of the dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.     

The Future Potential of Nuclear Power: Fission and Fusion

A comparison is made between fission and fusion as potential future power sources for mankind. Fission is seen to be less complex, but fusion is seen to be environmentally more benign.     

Low Radioactivity Fusion Reactor Based on the Spherical Tokamak with a Strong Magnetic Field

The most important feature of the spherical tokamak is the possibility of high-β plasma confinement (β is the ratio of plasma pressure to magnetic field pressure). So, spherical tokamak can be considered as a possible confinement system for D–³He fusion reactor. Present paper study the ability to develop powerful D–³He reactor based on a spherical tokamak (fusion power about 3 GW). The following parameters are considered as optimization criteria: (1) the ratio of confinement time to the value predicted by ITER98y2 scaling; (2) the neutron flux from the plasma.     

Evaluation of Decay Heat in Fusion Experimental Reactor

Decay heat in the blanket and shield of the Fusion Experimental Reactor (FER) was calculated using a newly developed radioactivation calculation code system THIDA-2. The decay heat at various time periods after shutdown were calculated. The decay heat level in the FER blanket was found to be at least one order of magnitude lower than in fission reactors at all time periods after shutdown. The necessity of following the transport of decay γ-rays in obtaining the detailed distribution of decay heat in the blanket was demonstrated. The validity of the γ-ray kerma factors used in the evaluation was also shown.